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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Assessment of lead tellurite glass for immobilizing electrochemical salt wastes from used nuclear fuel reprocessing

This paper provides an overview of research evaluating the use of lead tellurite glass as a waste form for salt wastes from electrochemical reprocessing of used nuclear fuel. The efficacy of using lead tellurite glass to immobilize three different salt compositions was evaluated: a LiCl-Li2O oxide reduction salt containing fission products from oxide fuel, a LiCl-KCl eutectic salt containing fission products from metallic fuel, and SrCl2. Physical and chemical properties of glasses made with these salts were characterized with X-ray diffraction, bulk density measurements, differential thermal analysis, chemical durability tests, scanning and transmission electron microscopies, and energy-dispersive X-ray spectroscopy. These glasses were found to accommodate high salt concentrations and have high densities, but further development is needed to improve chemical durability. (C) 2017 Published by Elsevier B.V.

Riley, Brian J.↗
DOE OSTI2019

Monitoring Aqueous Reprocessing Systems for Detection of Facility Misuse

Traditional approaches to material control and accounting at used nuclear fuel reprocessing facilities relies in large part on laboratory-based destructive assay of periodic samples taken from key stages. These techniques are highly accurate, but limited inspection frequency may limit the size and operation of a reprocessing facility. Online monitoring techniques using inline detectors can be applied to fuel reprocessing systems to improve the situational awareness as the facility is operating to provide additional indicators of the evolving state of the facility between normal inspection periods, potentially alerting regulators to the need for additional inspection. The Multi-Isotope Process (MIP) Monitor combines medium-resolution gamma detectors with statistical data analysis techniques to monitor reprocessing systems for anomalous conditions resulting from process control issues or malfeasance. Gamma spectra collected from the H-Canyon reprocessing facility were modeled using principal component analysis to demonstrate the efficacy of the MIP Monitor for detecting notional diversion scenarios.

safeguards, nuclear fuel reprocessing, gamma spect↗
DOE OSTI2024

Thermal Activation of Zirconium(IV) Acetylacetonate Catalysts to Enhance Polyurethane Synthesis and Reprocessing

Carbamate formation and exchange catalysts enable efficient polyurethane (PU) manufacturing, as well as emerging recycling and reprocessing methods for PU thermosets. Zirconium β-diketonate complexes, such as Zr acetylacetonate [Zr(acac) 4 ], are effective alternatives to toxic organotin catalysts that have been used for PU reprocessing. Here, we report that Zr(acac) 4 undergoes a thermally activated process in the PU network during reprocessing that transforms it into a more active carbamate exchange catalyst. This process is associated with the irreversible loss of acetylacetonate ligands and is not observed for the more sterically hindered Zr 2,2,6,6-tetramethyl-3,5-heptanedione [Zr(tmhd) 4 ] complex. Crossover experiments between PU thermoplastics indicated enhanced carbamate exchange after the thermal activation of Zr(acac) 4 in the presence of one of the PUs, whereas a sample of Zr(acac) 4 activated in the absence of the PU had no catalytic activity. Thermal gravimetric analysis suggested that this process is associated with the loss of one protonated acac ligand. Stress relaxation analysis of PU thermosets indicated a distinct change in the characteristic relaxation time associated with the thermal activation of Zr(acac) 4 at temperatures above 140 °C; no such change was observed for samples reprocessed using Zr(tmhd) 4 . Density functional theory and molecular experiments suggest that irreversible ligand exchange of acac with alkoxide or carbamate reduces the activation energy for urethane formation and reversion. Furthermore, the Zr(acac) 4 catalyst activated in the presence of a PU’s polyol precursor provided more porous and less dense PU foams compared to those made using the unactivated Zr(acac) 4 catalyst. Furthermore, these findings are important for developing improved PU synthesis and recycling processes. Thermally activating a catalyst during reprocessing may provide more nuanced control of the in-use and reprocessing characteristics of PU thermosets.

Alcohols↗
DOE OSTI2020

Capture of Iodine from Nuclear-Fuel-Reprocessing Off-Gas: Influence of Aging on a Reduced Silver Mordenite Adsorbent after Exposure to NO/NO 2

Iodine radioisotopes released during nuclear fuel reprocessing must be removed from the off-gas stream before discharge. One promising material for iodine capture is reduced silver mordenite (Ag 0 Z). Nevertheless, the adsorbent’s capacity will degrade, or age, over time when the material is exposed to other off-gas constituents. Though the overall impact of aging is known, the underlying physical and chemical processes are not. To examine these processes, Ag 0 Z samples were prepared and aged in 2% NO 2 in dry air and in 1% NO in N 2 gas streams at 150 °C for up to six months. Aged samples were then characterized using scanning electron microscopy, X-ray diffraction, Raman spectroscopy, and X-ray absorption spectroscopy. These techniques show that aging involves two overarching processes: (i) oxidation of the silver nanoparticles present in Ag 0 Z and (ii) migration of oxidized silver into the mordenite’s inner network. Silver on the nanoparticle’s surface is oxidized through adsorption of O 2 , NO, and NO 2 . Additionally, Raman spectroscopy and X-ray absorption spectroscopy indicate that nitrates are the primary products of this adsorption. Most of these nitrates migrate into the interior of the mordenite and exchange at framework binding sites, returning silver to its unreduced state (AgZ). The remaining nitrates exist at a persistent concentration without aggregating into bulk-phase AgNO 3 . X-ray absorption spectroscopy results further indicate that iodine adsorption occurs on not just Ag 0 Z but also on AgZ and a portion of the nitrates in the system. AgZ adsorbs a sizable quantity of iodine early in the aging process, but its capacity drops rapidly over time. For well-aged samples, nitrates are responsible for up to 95% of mordenite’s iodine capacity. Overall, these results have enhanced our understanding of the aging process in silver mordenite and are expected to guide the development of superior adsorbents for the capture of radioactive iodine from reprocessing off-gas.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Efficacy of a Solution-Based Approach for Making Sodalite Waste Forms for an Oxide Reduction Salt Utilized in the Reprocessing of Used Uranium Oxide Fuel

This paper describes various approaches for making sodalite with a LiCl-Li2O oxide reduction salt used to recover uranium from used oxide fuel. The approaches include sol-gel and solution-based synthesis processes. As-made products were mixed with 5 and 10 mass% of a Na2O-B2O3-SiO2 glass binder and these, along with product without a binder, were heated using either a cold-press-and-sinter method or hot uniaxial pressing. The results demonstrate the limitation of sodalite yield due to the fast intermediate reactions between Na+ and Cl- to form halite in solution and Li2O and SiO2 to form lithium silicates (e.g., Li2SiO3 or Li2Si2O5) in the calcined and sintered pellets. The results show that pellets can be made with high sodalite fractions (~92 mass%) and low porosities using a solution-based approach and this LiCl-Li2O salt.

electrochemical↗
DOE OSTI2020

Zinc-in-titania waste form for immobilizing lanthanide fission products from electrochemical reprocessing

This paper provides microstructural, microchemical, and bulk X-ray diffraction information for a zinc-in-titania waste form for immobilizing rare-earth oxide (REOx) fission products recovered through precipitation from electrochemical salt wastes. This waste form can be used to immobilize REOx at loadings as high as 40 mass% in a chemically durable matrix consisting of glass-bonded minerals including Zn2TiO4 as well as RE-containing phases of REOx, RE2Ti2O7, and REPO4 (monazite). Chemical durability data is also provided, and comparisons are drawn between this waste form and one made using a similar process with half the REOx loading.

rare earth, rare earth waste form, electrochemical↗
DOE OSTI2021

Light-Mediated Synthesis and Reprocessing of Dynamic Bottlebrush Elastomers under Ambient Conditions

We introduce a novel grafting-through polymerization strategy to synthesize dynamic bottlebrush polymers and elastomers in one step using light to construct a disulfide-containing backbone. The key starting material—α-lipoic acid (LA)—is commercially available, inexpensive, and biocompatible. When installed on the chain end(s) of poly(dimethylsiloxane) (PDMS), the cyclic disulfide unit derived from LA polymerizes under ultraviolet (UV) light in ambient conditions. Significantly, no additives such as initiator, solvent, or catalyst are required for efficient gelation. Formulations that include bis-LA-functionalized cross-linker yield bottlebrush elastomers with high gel fractions (83–98%) and tunable, supersoft shear moduli in the ~20–200 kPa range. An added advantage of these materials is the dynamic disulfide bonds along each bottlebrush backbone, which allow for light-mediated self-healing and on-demand chemical degradation. These results highlight the potential of simple and scalable synthetic routes to generate unique bottlebrush polymers and elastomers based on PDMS.

36 MATERIALS SCIENCE↗
DOE OSTI2018

New AREVA's Investment in Continuous Improvement to Increase Efficiency of La Hague Reprocessing Lines - 18588

New AREVA currently operates 2 commercial reprocessing lines at their La Hague facility in France. Each line is capable of reprocessing up to 800 tHM/year of commercial spent light water reactor fuel. The site and process is self-contained with all waste arising from the reprocessing operations segregated into various waste streams and stabilized at the site. New AREVA operates a culture of continuous improvement and continues to invest in new technology, processes, waste management and procedures that have increased plant throughput while reducing waste volumes, operating cost while at the same time maintaining our strong safety culture. New investments have also been driven by changing customer needs and new markets. A good example of this has emerged over the last several years as New AREVA has been transporting a large amount of Research Reactor (RR) Uranium-Aluminum (UAl) type spent fuel (SF) on behalf of international RR operators to the La Hague site. Last year more than 100 RR fuel elements shipments were carried out by New AREVA (for fresh and spent fuel). In our current commercial lines the reprocessing of RRSF has been limited to approximately 1.4 ton a year. To meet increasing demand for reprocessing of RRSF, the range of RRSF that can now be reprocessed at La Hague has been extended to include Uranium-Aluminium-Silicium fuel (USi also called Silicide fuel) and also non-intact RRSF. In addition, New AREVA supported large scale research, development and engineering programs to identify processes/technology to allow the La Hague Facilities to reprocess a targeted average of additional 25 tons/year of non-commercial fuel types (RRSF and also fast reactor fuel, specific MOX fuel... ). A new facility head end called Traitement de Combustibles Particuliers (TCP) facility is currently planned to be implemented in the La Hague plant allowing reprocessed special fuels to be introduced into the existing commercial processing lines. The purpose of the paper is to demonstrate New AREVA's culture of continuous improvement that has been implemented at our commercial reprocessing facility highlighting how we have invested in processes and procedures for improving services to the nuclear operators. A case study will be presented related to our latest development of the TCP head end that will allow New AREVA to reprocess a targeted average of additional 25 tons/year of specialized fuels. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2018

TRMM Version 8 Reprocessing Improvements and Incorporation into the GPM Data Suite

The National Aeronautics and Space Administration (NASA) has always included data reprocessing as a major component of every science mission. A final reprocessing is typically a part of mission closeout (known as phase F).The Tropical Rainfall Measuring Mission (TRMM) is currently in phase F, and NASA is preparing for the last reprocessing of all the TRMM precipitation data as part of the closeout. This reprocessing includes improvements in calibration of both the TRMM Microwave Imager (TMI) and the TRMM Precipitation Radar (PR). An initial step in the version 8 reprocessing is the improvement of geolocation. The PR calibration is being updated by the Japan Aerospace Exploration Agency (JAXA) using data collected as part of the calibration of the Dual-Frequency Precipitation Radar (DPR) on the Global Precipitation Measurement (GPM) Core Observatory. JAXA undertook a major effort to ensure TRMM PR and GPM Ku-band calibration is consistent. A major component of the TRMM version 8 reprocessing is to create consistent retrievals with the GPM version 05 (V05) retrievals. To this end, the TRMM version 8 reprocessing uses retrieval algorithms based on the GPM V05 algorithms. This approach ensures consistent retrievals from December 1997 (the beginning of TRMM) through the current ongoing GPM retrievals. An outcome of this reprocessing is the incorporation of TRMM data products into the GPM data suite. Incorporation also means that GPM file naming conventions and reprocessed TRMM data carry the V05 data product version. This paper describes the TRMM version 8 reprocessing, focusing on the improvements in TMI level 1 products.

Stocker, E. F.↗
DOE OSTI2016

Tritium Distribution and Capture Requirements for UNF Reprocessing

Four radionuclides have been identified as being sufficiently volatile in reprocessing of nuclear fuel that their gaseous release needs to be controlled to meet U.S. regulatory requirements. These radionuclides are {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I. Previous studies have discussed the need for treating off-gas streams to control {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I from the reprocessing of used nuclear fuel (UNF) and the regulations that set the limits for releases from such a facility. These regulations ultimately prescribe the decontamination factor (DF) that must be met in the facility. Jubin et al. examined the role of off-gas pathways since the gaseous radionuclides can partition to different reprocessing off-gas streams. Based on this report and subsequent analysis, the required DF for {sup 129}I could be 1000 to 3000. The required DF for {sup 3}H could be as high as 200, depending on the age of the fuel processed. The DF for {sup 85}Kr could be up to ∼35, also depending on fuel age. The required DF for {sup 14}C is 1 in many cases (no treatment required) but could be higher in specific cases. While iodine has been the focus of a number of studies due to the high DF requirements on multiple escape pathways, it is clear from the studies noted above that the control of tritium will also be a significant issue for future reprocessing plants, especially if processing relatively short-cooled fuels (<10 yr post reactor discharge). Tritium contained in a UNF matrix will be released into the liquid and off-gas streams during fuel reprocessing. The amount released into the process streams is a function of the type and burnup of the fuel processed, and the chemical processes employed within the reprocessing plant. Previous studies have examined the tritium capture requirements for the reprocessing facility assuming that all of the tritium is volatilized in the head-end operations, i.e., tritium pretreatment is employed and tritium is completely released from the cladding. This is clearly a worst case scenario from the standpoint of the amount of tritium that could be released, but the scenario tends to limit the release pathway to the head-end off-gas stream. In this analysis, the volume of HTO and the HTO to H{sub 2}O ratio for each of the three plant-wide tritium DFs were determined. The overall amount of HTO that had to be captured from the off-gas streams was virtually the same for the three DF cases evaluated; 98%, 99% and 99.5% of the tritium in the fuel entering the plant needed to be recovered for overall plant DFs of 50, 100 and 200, respectively. The use of the 'tritium barrier' to limit the spread of HTO from the head-end systems to the balance of the plant results in an increase in tritiated waste concentration in the head-end recycle acid systems which results in a decrease in the total amount of HTO + H{sub 2}O that must recovered from the gaseous effluent from the plant. The use of the 'tritium barrier' decreases the waste volume due to the higher HTO:H{sub 2}O ratio in the treated off-gas stream. The results also show that increasing the acid recycle percentage increases the concentration of HTO within the UNF reprocessing plant. This is then manifested in less HTO + H{sub 2}O that must be recovered from the gaseous effluents. Finally, the application of tritium recovery on the recycle acid reduces the HTO concentration in the recycle acid stream and generates a separate HTO + H{sub 2}O stream that must be managed. However, the resulting volume of tritium-containing waste generated plant-wide is a function of the selectivity of the tritium recovery system and the ability of the tritium recovery system to limit the accumulation of HTO in the recycle acid such that the HTO + H{sub 2}O recovered from gaseous effluents does not contribute significantly to the overall tritium waste volume. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Effect of reprocessing cycles on the degradation of polypropylene copolymer filled with talc or montmorillonite during injection molding process

Mechanical recycling of polymeric materials is a favorable technique resulting in economic and environmental benefits, especially in the case of polymers with a high production volume as the polypropylene copolymer (PP). However, recycling by reprocessing techniques can lead to thermal, mechanical or thermo-oxidative degradation that can affect the structure of the polymer and subsequently the material properties. PP filled with montmorillonite (MMT) or talc are widely produced and studied, however, its degradation reactions by reprocessing cycles are poorly studied so far. In this study, the effects of reprocessing cycles in the structure and in the properties of the PP/MMT and PP/Talc were evaluated. The samples were mixed with 5% talc or MMT Cloisite C15A in a twin-screw extrusion. After extrusion, this filled material was submitted to five reprocessing cycles through an injection molding process. In order to evaluate the changes induced by reprocessing techniques, the samples were characterized by DSC, FT-IR, Izod impact and tensile strength tests. The study showed that Young modulus, elongation at brake and Izod impact were not affected by reprocessing cycles, except when using talc. In this case, the elongation at brake reduced until the fourth cycle, showing rigidity increase. The DSC results showed that melting and crystallization temperature were not affected. A comparison of FT-IR spectra of the reprocessed indicated that in both samples, between the first and the fifth cycle, no noticeable change has occurred. Thus, there is no evidence of thermo oxidative degradation. In general, these results suggest that PP reprocessing cycles using MMT or talc does not change the material properties until the fifth cycle.

36 MATERIALS SCIENCE↗
DOE OSTI2026

Recyclability of reversible polymer networks over a Dozen reprocessing cycles

Reversible polymer networks capable of reversible reactions along their backbone provide a promising strategy for addressing waste management challenges associated with conventional thermoset products. However, reversible polymer networks cannot be recycled infinitely. Unavoidable side reactions eventually cause significant degradation of mechanical properties or loss of recyclability. This study aims to answer a simple yet critical question on reversible polymer network systems: how many reprocessing cycles can be achieved before profound mechanical degradation or a loss of recyclability? The recyclability of Diels–Alder (DA) network samples was evaluated by repeating consistent reprocessing cycles until they were no longer reprocessable. To enhance their recyclability, the chemical structure of the maleimide precursor was tailored to mitigating side reactions by maleimide homopolymerization. Remarkably, by employing a maleimide precursor with alkyl substitutions on its phenyl group, the DA network system attained 12 reprocessing cycles using injection molding at 160 °C without significant degradation of mechanical properties. However, the 13th reprocessing cycle did not succeed. Rheological analysis revealed the accumulation of non-reversible bonds within the network structure during repeated reprocessing, despite fairly consistent mechanical properties under operating conditions. This study demonstrates that the rational design of the maleimide precursor is an effective means to enhance the reprocessability of DA networks.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2014

Radioactive Semivolatiles in Nuclear Fuel Reprocessing

In nuclear fuel reprocessing, various radioactive elements enter the gas phase from the unit operations found in the reprocessing facility. In previous reports, the pathways and required removal were discussed for four radionuclides known to be volatile, 14 C, 3 H, 129 I, and 85 Kr. Other, less volatile isotopes can also report to the off-gas streams in a reprocessing facility. These were reported to be isotopes of Cs, Cd, Ru, Sb, Tc, and Te. In this report, an effort is made to determine which, if any, of 24 semivolatile radionuclides could be released from a reprocessing plant and, if so, what would be the likely quantities released. As part of this study of semivolatile elements, the amount of each generated during fission is included as part of the assessment for the need to control their emission. Also included in this study is the assessment of the cooling time (time out of reactor) before the fuel is processed. This aspect is important for the short-lived isotopes shown in the list, especially for cooling times approaching 10 y. The approach taken in this study was to determine if semivolatile radionuclides need to be included in a list of gas-phase radionuclides that might need to be removed to meet Environmental Protection Agency (EPA) and Nuclear Regulatory Commission (NRC) regulations. A list of possible elements was developed through a literature search and through knowledge and literature on the chemical processes in typical aqueous processing of nuclear fuels. A long list of possible radionuclides present in irradiated fuel was generated and then trimmed by considering isotope half-life and calculating the dose from each to a maximum exposed individual with the US EPA airborne radiological dispersion and risk assessment code CAP88 (Rosnick 1992) to yield a short list of elements that actually need to be considered for control because they require high decontamination factors to meet a reasonable fraction of the regulated release. Each of these elements is then discussed with respect to what is known in the literature about their behavior in a reprocessing facility. The context for the evaluation in this document is a UO2-based fuel processed through an aqueous-based reprocessing system with a TBP-based solvent extraction chemistry. None of these elements form sufficiently volatile compounds in the context of the reprocessing facility to be of regulatory concern.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗